ASTM E2005-21

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Designation: E2005 − 10 (Reapproved 2015) E2005 − 21
Standard Guide for
Benchmark Testing of Reactor Dosimetry in Standard and
Reference Neutron Fields
This standard is issued under the fixed designation E2005; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope
1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of
the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of
certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement
consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to
benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in
derived neutron dosimetry results.
1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility
of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of
regulatory limitations prior to use.
1.4 This international standard was developed in accordance with internationally recognized principles on standardization
established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued
by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
2. Referenced Documents
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Dosimetry
E261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques
E263 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron
E264 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel
E265 Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32
E266 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum
E343 Test Method for Measuring Reaction Rates by Analysis of Molybdenum-99 Radioactivity From Fission Dosimeters
(Withdrawn 2002)
E393 Test Method for Measuring Reaction Rates by Analysis of Barium-140 From Fission Dosimeters
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance
E523 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.05 on Nuclear
Radiation Metrology.
Current edition approved Oct. 1, 2015Feb. 1, 2021. Published November 2015March 2021. Originally approved in 1999. Last previous edition approved in 20102015 as
E2005 - 10.E2005– 10(2015). DOI: 10.1520/E2005-10R15.10.1520/E2005-21.
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
The last approved version of this historical standard is referenced on www.astm.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E2005 − 21
E526 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Titanium
E704 Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238
E705 Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
E1297 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Niobium
E2006 Guide for Benchmark Testing of Light Water Reactor Calculations
3. Significance and Use
3.1 This guide describes approaches for using neutron fields with well known characteristics to perform calibrations of neutron
sensors, to intercompare different methods of dosimetry, and to corroborate procedures used to derive neutron field information
from measurements of neutron sensor response.
3.2 This guide discusses only selected standard and reference neutron fields which are appropriate for benchmark testing of
light-water reactor dosimetry. The Standard Fields considered are here include neutron source environments that closely
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approximateapproximate: a) the unscattered neutron spectra from Cf spontaneous fissionfission; and b) the U thermal neutron
induced fission. These standard fields were chosen for their spectral similarity to the high energy region (E > 2 MeV) of reactor
spectra. The various categories of benchmark fields are defined in Terminology E170.
3.3 There are other well known neutron fields that have been designed to mockup special environments, such as pressure vessel
mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel.” When such mockups
are suitably characterized, they are also referred to as benchmark fields. A variety of these engineering benchmark fields have been
developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. These special
benchmark experiments are discussed in Guide E2006, and in Refs (1) and (2).
4. Neutron Field Benchmarking
4.1 To accomplish neutron field “benchmarking,” one must perform irradiations in a well-characterized neutron environment, with
the required level of accuracy established by a sufficient quantity and quality of results supported by a rigorous uncertainty
analysis. What constitutes sufficient results and their required accuracy level frequently depends upon the situation. For example:
4.1.1 Benchmarking to test the capabilities of a new dosimeter;
4.1.2 Benchmarking to ensure long-term stability, or continuity, of procedures that are influenced by changes of personnel and
equipment;
4.1.3 Benchmarking measurements that will serve as the basis of intercomparison of results from different laboratories;
4.1.4 Benchmarking to determine the accuracy of newly established benchmark fields; and
4.1.5 Benchmarking to validate certain ASTM standard methods or practices which derive exposure parameters (for example,
fluence > 1 MeV or dpa) from dosimetry measurements and calculations.
5. Description of Standard and Reference Fields
5.1 There are a few facilities which can provide certified “free field” fluence irradiations. The following provides a list of such
facilities. The emphasis is on facilities that have a long-lived commitment to development, maintenance, research, and international
interlaboratory comparison calibrations. As such, discussion is limited to recently existing facilities.
5.2 Cf Fission Spectrum—Standard Neutron Field:
5.2.1 The standard fission-spectrum fluence from a suitably encapsulated Cf source is characterized by its source strength, the
distance from the source, and the irradiation time. In the U.S., neutron source emission rate calibrations are all referenced to source
calibrations at the National Institute of Standards and Technology (NIST) accomplished by the MnSO technique (3). Corrections
The boldface numbers given in parentheses refer to a list of references at the end of the text.
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for neutron absorption, scattering, and other than point-geometry conditions may, by careful experimental design, be held to less
than 3 %. Associated uncertainties for the NIST Cf irradiation facility are discussed in Ref (4). The principal uncertainties,
which only total about 2.5 %, come from the source strength determination, scattering corrections, and distance measurements.
Extensive details of standard field characteristics and values of measured and calculated spectrum-averaged cross sections are all
given in a compendium, see Ref (5).
5.2.2 The NIST Cf sources have a very nearly unperturbed spontaneous fission spectrum, because of the light-weight
encapsulations, fabricated at the Oak Ridge National Laboratory (ORNL), see Ref (6).
5.2.3 For a comprehensive view of the calibration and use of a special (32 mg) Cf source employed to measure the
spectrum-averaged cross section of the Nb(n,n') reaction, see Ref (7).
5.3 U Fission Spectrum—Standard Neutron Field:
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5.3.1 Because U fission is the principal source of neutrons in present nuclear reactors, the U fission spectrum is a
fundamental neutron field for benchmark referencing or dosimetry accomplished in reactor environments. This remains true even
for low-enrichment cores which have up to 30 % burnup.
5.3.2 There are currently two U standard fission spectrum facilities, one in the thermal column of the NIST Research Reactor
(8) and one at CEN/SCK, Mol, Belgium (9).
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5.3.3 A standard U neutron field is obtained by driving (fissioning) U in a field of thermal neutrons. Therefore, the fluence
rate depends upon the power level of the driving reactor, which is frequently not well known or particularly stable. Time dependent
fluence rate, or total fluence, monitoring is necessary in the U field. Certified fluence irradiations are monitored with the
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Ni(n,p) Co activation reaction. The fluence-monitor calibration must be benchmarked.
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5.3.4 For U, as for Cf irradiations, small (nominally < 3 %) scattering and absorption corrections are necessary. In addition,
for U, gradient corrections of the measured fluence which do not simply depend upon distance are necessary. The scattering and
gradient corrections are determined by Monte Carlo calculations. Field characteristics of the NIST U Fission Spectrum Facility
and associated measured and calculated cross sections are given in Ref (5).
5.4 There are several additional facilities that can provide free field fluence irradiations that qualify as reference fields.fields (10,
11). The following is a list of some of the facilities that have characterized reference fields:
5.4.1 Annular Core Research Reactor (ACRR) Central Cavity – Reference Neutron Field (1012, 1113),
5.4.2 ACRR Lead-Boron Cavity Insert – Reference Neutron Field (1113),
5.4.3 YAYOI fast neutron field – Reference Neutron Field (1214, 1315),, 16) [no longer operational],
5.4.4 SIGMA-SIGMA neutron field – Reference Neutron Field (1214, 1315).) [no longer operational],
5.4.5 LR0-Rez neutron field – Reference Neutron Field (10),
5.4.6 TRIGA-JSI neutron field – Reference Neutron Field (10).
6. Applications of Benchmark Fields
6.1 Notation—Reaction Rate, Fluence Rate, and Fluence—The notation employed in this section will follow that in E261
(Standard Practice for Determining Neutron Fluence Rate, and Spectra by Radioactivation Techniques) except as noted. The
reaction rate, R, for some neutron-nuclear reaction {reactions/[(dosimeter target nucleus)(second)]} is given by:
`
R 5 σ E φ E dE (1)
* ~ ! ~ !
o
or:
R 5 σ¯ φ (2)
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where:
–24 2
σ(E) = the dosimeter reaction cross section at energy E (typically of the order of 10 cm ),
φ(E) = the differential neutron fluence rate, that is the fluence per unit time and unit energy for neutrons with energies between
–2 –1 –1
E and E + dE (neutrons cm s MeV ),
–2 –1
φ = the total fluence rate (neutrons cm s ), the integral of φ(E) over all E, and
σ¯ = the spectral-averaged value of σ(E), R/φ.
σ¯ = the spectral-averaged value of σ(E), that is, R/φ.
NOTE 1—Neutron fluence and fluence rate are defined formally in Terminology E170 under the listing “particle fluence.” Fluence is just the time integral
of the fluence rate over the time interval of interest. The fluence rate is also called the flux or flux density in many papers and books on neutron transport
theory.
6.1.1 The reaction rate is found experimentally using an active instrument such as a fission chamber (see Ref (1417)) or a passive
dosimeter such as a solid state track recorder (see Test Method E854), a helium accumulation fluence monitor (see Test Method
E910), or a radioactivation dosimeter (see Practice E261). For the radioactivation method, there are also separate standards for
many particularly important dosimetry nuclides, for example, see Test Methods E263, E264, E265, E266, E343, E393, E523, E526,
E704, E705, and E1297.
6.2 Fluence Rate Transfer: Note that if one determines φ = R/σ¯ from Eq 2, then the uncertainty in φ will be a propagation of the
uncertainties in both R and σ¯. The uncertainty in σ¯ is frequently large, leading to a less accurate determination of φ than desired.
However, if one can make an additional irradiation of the same type of dosimeter in a standard neutron field with known fluence
rate, then one may apply Eq 2 to both irradiations and write
φ 5 φ ~R /R ! ~σ¯ /σ¯ ! (3)
A B A B B A
where “A” denotes the field of interest and “B” denotes the standard neutron field benchmark. In Eq 3 the ratios of spectral
average cross section, σ¯ /σ¯ will have a small uncertainty if the spectral shapes φ (E) and φ (E) are fairly similar. There may
A B A B
also be important cancellation of poorly known factors in the ratio R /R , which will contribute to the better accuracy of Eq 3.
A B
Whether φ is better determined by Eq 3 or Eq 2 must be evaluated on a case by case basis. Often the fluence rate from Eq 3 is
substantially more accurate and provides a very useful validation of other dosimetry. The use of a benchmark neutron field
irradiation and Eq 3 is called fluence rate transfer.
6.2.1 Certified Fluence or Fluence Rate Irradiations—The primary benefit from carefully-made irradiations in a standard neutron
field is that of knowing the neutron fluence rate. Consider the case of a lightly encapsulated Cf sintered-oxide bead, which has
an emission rate known to about 61.5 % by calibration in a manganese bath (MnSO solution). Further, consider a dosimeter pair
irradiated in compensated beam geometry (with each member of the pair equidistant from, and on opposite sides of, the Cf
source). For such an irradiation in a large room (where very little room return occurs), the fluence rate – with a Cf fission
spectrum – is known to within 63 % from the source strength, and the average distance of the dosimeter pair from the center of
the source. Questions concerning in- and out-scattering by source encapsulation, source and foil holders, and foil thicknesses may
be accurately investigated by Monte Carlo calculations. There is no other neutron-irradiation situation that can approach this level
of accuracy in determination of the fluence or fluence rate.
6.2.2 Fluence Transfer Calibrations of Reference Fields—The benefit of irradiating with a source of known emission rate is lost
when one must consider reactor cores or, even, thermal-neutron fissioned U sources. When the latter are carefully constructed
to provide for an unmoderated U fission neutron spectrum, this mentioned disadvantage can be circumvented by a process called
fluence transfer. As explained briefly in 6.2, this process is basically as follows. A gamma-counter (spectrometer) geometry is
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chosen to enable proper counting of the activities of a particular isotopic reaction for example, Ni(n,p) Co, after irradiation in
252 235 252
either a Cf or U field. Then the Cf irradiation is accomplished and the nickel foil counted. From this, a ratio of the
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dosimeter response divided by the Cf certified fluence is determined. Subsequently, an identical nickel is irradiated in the U
spectrum and that foil is counted with the same counter geometry. Within the knowledge of the ratio of the spectrum average
spectrum-averaged cross sections in the two spectra, knowledge of the counter response to the recent irradiation yields the average
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U fluence. Note, the average fluence is measured. The thermal fluence rate at the U sources may not have been constant over
the time of the irradiation but that time is assumed to be short relative to the 70 day half-life of the Co, which monitors the fast
neutron fluence through-out the irradiation. The method of calibration is termed fluence rate transfer because it is fluence rate
which is determined, and there is no need to determine the absolute radioactivity of the dosimeters. Relative response of the same
counter geometry is the only requirement.
6.2.3 Reactor Irradiations—In principle, the same fluence-transfer procedures can be applied to more complex irradiations.
However, there are certain other situations which must be considered and weighed to determine if fluence transfer or reaction rate
determination is the better method. Also, remember that error estimation can be examined by using both methods.
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6.2.3.1 If radioactivation dosimeters are employed for long term irradiations in a power reactor, the fluence at a dosimeter location
can be
...